The sloshing in a confined tank is essential to evaluate the safety function of floating plants, e.g., offshore floating nuclear power plants (OFNPs). The transient plant behavior has been simulated with nuclear system analysis codes. In turn, the sloshing motion has been simulated with computational fluid dynamics (CFD) codes. Although some system codes can calculate three-dimensional flow, the available features were limited, exceptionally dynamic phasic behaviors, including the sloshing motions. The paper addresses the time-dependent acceleration as body force in the momentum equation of the system code TRACE and the validation of the pressure impact acting on the wall and fixed-roof of the cylindrical tank. The modified TRACE code was validated against the sloshing experiment. In addition, the TRACE code results were compared with CFD code (Star-CCM+) results. Two different methods were compared: the volume of fluid (VOF) method and the two-fluid flow method, namely the Eulerian multiphase model (EMP). The results of the flow models indicate that the free-surface VOF model agrees with the experimental results. However, the fast transient motions are suppressed for the two-flow model in the CFD and TRACE code.

In this study, we developed a comb-shaped microfluidic device that can efficiently trap and culture a single cell (bacterium). Conventional culture devices have difficulty in trapping a single bacterium and often use a centrifuge to push the bacterium into the channel. The device developed in this study can store bacteria in almost all growth channels using the flowing fluid. In addition, chemical replacement can be performed in a few seconds, making this device suitable for culture experiments with resistant bacteria. The storage efficiency of microbeads that mimic bacteria was significantly improved from 0.2% to 84%. We used simulations to investigate the pressure loss in the growth channel. The pressure in the growth channel of the conventional device was more than 1400 PaG, whereas that of the new device was less than 400 PaG. Our microfluidic device was easily fabricated by a soft microelectromechanical systems method. The device was highly versatile and can be applied to various bacteria, such as Salmonella enterica serovar Typhimurium and Staphylococcus aureus.

The efficient synthesis of amino acid Schiff base copper(II) complexes using a microfluidic device was successfully achieved. Schiff bases and their complexes are remarkable compounds due to their high biological activity and catalytic function. Conventionally, products are synthesized under reaction conditions of 40 °C for 4 h using a beaker-based method. However, in this paper, we propose using a microfluidic channel to enable quasi-instantaneous synthesis at room temperature (23 °C). The products were characterized using UV–Vis, FT–IR, and MS spectroscopy. The efficient generation of compounds using microfluidic channels has the potential to significantly contribute to the efficiency of drug discovery and material development due to high reactivity.

To understand the eutectic reaction mechanism and the relocation behavior of the core debris is indispensable for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast re-actors (SFRs). This paper addresses reaction products and their distribution of the eutectic melting/so-lidifying reaction of boron carbide (B4C) and stainless-steel (SS). The influence of the existence of carbon on the B4C-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron micro-scopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the B4C-SS samples, a new layer was formed between B4C/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or Fe2B. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples. All samples observed the Cr-rich domain and Fe and Ni-rich domain after the re-action. These domains might be formed during the solidifying process.(c) 2022 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

In boiling water reactors (BWRs), the fuel assembly incorporates nonuniform elements such as part-length fuel rods and water rods capable of modifying the moderator density distribution and axial pressure drop in order to improve the economic efficiency and the thermal margin. This study focuses on the void fraction distribution in a rod bundle with different nonheated rod arrangements in order to evaluate the effect of water rods as a nonheated section on the boiling flow dynamics in the rod bundle. A boiling flow experiment was conducted using the test facility for 3D thermal hydraulics in light water reactors (SIRIUS-3D) and a linear accelerator-driven high-energy X-ray computed tomography (CT) system under BWR rated pressure conditions. The test section was a 5 × 5 rod bundle of heated length 3708 mm that simulated a BWR rod bundle. The void fraction distribution in the 5 × 5 rod bundle was acquired at six heights via X-ray CT imaging. The experimental results show the effect of nonheated rod arrangements on the local void fraction and the evolution of boiling flow in the rod bundle.

Disruption in core coolant injection leads to pool boiling in a reactor. The void behavior during this pool boiling significantly affects long-term core cooling. Thus, a pool boiling experiment was performed using a 5 x 5 heated rod bundle with steady-state and sinusoidal oscillating thermal power at 0.05 and 0.10 Hz for a pressure range of 0.5-7.2 MPa. Subchannel void sensors were installed in the 5 x 5 rod bundle. They partially simulated the fuel assembly of an actual reactor core to analyze the spatio-temporal variations and frequency characteristics of the temporal evolution of void fraction distribution. Analysis of experimental results with wavelet transforms revealed the two-phase flow structure due to thermal power oscillation and its sensitivity to pressure and fre-quency of thermal power oscillation.

Steam explosions are considered industrial disasters in many industries, including the metal and paper industries. Steam explosions are initiated by the collapse of vapor film, which separates hot molten material and water. We have proposed the countermeasure by adding polyethylene oxide (PEO) into the water to stabilize the vapor film. A molten tin jet was immersed in these PEO aqueous solutions for a wide range of molecular weight (400-8000 kg/mol) and concentration of PEO (up to 0.1 wt%) to identify the occurrence of steam explosion. The steam explosion was suppressed for higher molecular weight and denser concentrations of PEO. The measured cloud-point temperature and solid-sphere quenching temperature indicate that steam explosion can be suppressed when the cloud-point temperature is a few Kelvins below the saturation temperature so that the solute PEO is deposited near the vapor-liquid interface to increase the viscosity more than two thousand times higher than that of water to stabilize the vapor film as a steam explosion retardant. For the lower molecular weight of PEO, the cloud-point temperature is close to the saturation temperature, PEO concentration must be thicker to precipitate the PEO to stabilize the vapor film.

The suppression measures of steam explosions are essential for the integrity of a containment vessel in light water reactors during severe accidents. Polyethylene glycol (PEG) can potentially suppress steam explosions when a molten metal falls into a PEG aqueous solution pool. We also reported that PEG is a reliable additive to suppress spontaneous steam explosions with and without external triggers. Water with a PEG becomes cloudy when the solution temperature exceeds a specific limit, known as the cloud point. The dissolved PEG starts precipitating beyond the cloud point temperature. Such precipitates stabilize a vapor film around a high-temperature molten metal and prevent the fine mixing of molten metal in the solution. To evaluate the suppressive effect of PEG solution and the controllability of steam explosion, a small-scale experiment was conducted in which a prototypic reactor metal was melted and released into a solution pool. Type-304 stainless steel and 30 wt% Zr mixed with type-304 stainless steel were used as test metals. The test metal was melted by induction heating in a crucible and kept at 1700°C. The molten metal jet was immersed in a solution pool of 20°C. Visual observation of the interaction between the molten metal and solution shows the effectiveness of steam explosion retardant and the required PEG solution concentration for a molecular weight of 4 million grams per mol.

The molten-core-coolant interaction is important in assessing the integrity of a reactor pressure vessel (RPV) and containment building (CB). In case of RPV failure during in-vessel retention (IVR), the breakage of the RPV will most likely occur at the level of the upper metallic layer due to the focusing effect. A possible steam explosion may result from the interaction between the metallic melt and water in the CB. If the metallic melt, composed of steel mixed with metallic species such as zirconium and uranium, undergoes oxidation during the premixing, triggering, and explosion phases, the melt oxidation influences the progress of the steam explosion. In this study, a small-scale experiment was conducted by dropping molten droplets composed of stainless steel mixed with zirconium into a water pool. The oxidation characteristics in water, such as drop oxidation, oxide film thickness, and element mapping of the solidified drops, were evaluated by an oxygen analyzer and Scanning Electron Microscope-Energy Dispersive X-ray Spectrometry (SEM-EDX) to clarify the effect of the metal composition of zirconium and stainless steel.

A critical heat flux (CHF) under forced boiling flow conditions is essential for the thermal design of the reactor core in light water reactors. An optical fiber temperature sensor can measure temperature distribution with high temporal and spatial resolutions and contribute to the CHF heat transfer characteristics. This study focuses on the development of the rod surface temperature measurement technique using optical fiber sensors in a heated rod bundle at high pressure and temperature. A heater rod with an optical fiber sensor mounted on its surface was developed. A forced boiling flow experiment was conducted at high pressure of up to 7.2 MPa using a 2 × 2 rod bundle assembled from the heater rods. The rod surface temperature throughout the heated section was measured at 2.6 mm intervals in the axial direction at the sampling rate of 100 Hz. The measurement results show that the local and instantaneous temperature fluctuations of the heated rod, i.e., local drying and rewetting, occurred intermittently near the CHF condition.

In case of severe accidents in water-cooled reactors, the molten core must be cooled down readily and stably. Injecting water is an efficient measure to cool down the molten core, while steam explosion threatens vessel integrity. We have proposed the steam explosion retardant, which is an aqueous solution of polyethylene oxide (PEO) since it stabilizes the vapor film, which separates hot material and water. The drawback of vapor film stability becomes significant for long-term cooling, as the film-boiling heat transfer gives a lower heat transfer rate. Quenching experiments were conducted to investigate the vapor-film collapse temperature where the boiling regime shifts from film boing to nucleate boiling. A 30 mm stainless-steel sphere at 700 oC was immersed into the solution pool. The vapor film immediately collapsed after immersion in a low-temperature (below 45 oC) water pool. The quenching temperature reduces steeply with increasing water pool temperature. The vapor film stabilized below 300 oC for 0.1wt% PEO solution, which indicates the sufficient performance of steam explosion retardant. As the water temperature rose, the vapor film collapse temperature became lower. The collapse temperature reverted to around 90 oC between the water and the PEO solution. Therefore, the steam explosion retardant suppresses the steam explosion at the early stage, while long-term coolability is better than water, as the vapor film collapses at higher temperatures.

An experimental campaign was conducted to determine drag coefficients of circular cylinders in axial flow of water for a wide range of length-to-diameter ratios (L/D) from 2 to 35. Drag force acting on the circular cylinders was acquired at a velocity of 2 to 6 m/s, which corresponds to the Reynolds numbers of 2.2 x 10(5) to 7.0 x 10(5). The experimental data were validated by analysis using the commercial CFD code, Star-CCM+. The measured drag coefficients increased monotonically as the L/D increased with the range of Reynolds numbers, and were almost constant regardless of the Reynolds number. The drag coefficient was decomposed by the term of form drag and skin friction. As the value of L/D increased, the drag coefficient increased due to the skin friction drag term. The drag coefficient correlation was proposed as a function of L/D. The predicted values were consistent with both the experimental data and numerical simulation results. Furthermore, this correlation was applied to a fuel assembly with a channel box, it was shown that the drag coefficient can be predicted for a wide range of Reynolds numbers by specifying an appropriate form drag term.

In general, digital twin refers to a digital replica of physical process or systems. We have proposed the digital twin concept of a mother design for twin children (experiment and simulation) with a help of the additive manufacturing technology. Moreover, the digital triplet concept to derive the regressive and well-correlated design on the basis of knowledge and experiences with a help of machine learning and statistics. The paper addresses our developed technologies of powder production, powder metallurgy, three-dimensional modelling, additive manufacturing to expand material variations for broadening the application of additive manufacturing. Devised additive-manufacturing method is devoted for complex structures with scalable measurement and control systems, including wide variety of thermal-hydraulic applications together with computational multi-fluid dynamic simulations in terms of the nuclear safety.

In a hypothetical severe accident scenario in light water reactors (LWR), after the loss of primary coolant, the melted core moves to the lower plenum of the reactor pressure vessel and accumulate there. The melted core cautiously releases the decay heat which forms a pool of melted core, called corium, and turbulent natural convection starts. The decay heat transferred by the corium to the vessel may result in vessel failure due to the focusing effect in the absence of an effective cooling system. A numerical analysis is carried out using STAR-CCM+ commercial software to investigate the ability of the existing turbulence models and Algebraic Heat Flux Model (AHFM) to simulate the natural convective heat transfer phenomena in the corium pools. Efforts have been made to predict the BALI experimental results which are designed to simulate the heat transfer in corium pools, in the framework of severe accident studies. For BALI experiments various test campaigns are run varying the internal Rayleigh number, the viscosity of the simulant fluid, and test facility height. Various such experimental cases are verified and compared with numerical simulations. Temperature profiles along a vertical line, surface heat flux along the curved surface, average Nusselt number on the top, and curved surfaces are estimated. The effect of wall boundary conditions on heat transfer is analyzed. The numerical analysis is extended to Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) geometries. The CFD analysis predicts the stratified zone and upper mixing zone very well, and the temperature profiles and average heat transfer values are in agreement with the experiments.

Borated water is injected into boiling-water reactors as a neutron absorber, and it is, therefore, essential to understand the effect of the solution concentration on its diffusion. We have developed a technique for measuring concentration distributions using a microchannel and near-infrared imaging. This measurement technique was applied to borated water in the present study, precisely an aqueous solution of sodium pentaborate. Laminar flows of water and this aqueous solution with 0.12 mol/L, 0.24 mol/L, and 0.32 mol/L at temperatures from 30 degrees C to 70 degrees C were mixed in a microchannel at Reynolds numbers of less than 2, and images were acquired at a wavelength of 1900 nm. From the obtained images, the concentration distribution was visualized, and the diffusion coefficient was estimated as being between 3.6 x 10(-10) m(2) /s and 9.8 x 10(-10) m(2)/s. The diffusion coefficient was correlated positively with both temperature and concentration. (C) 2022 Elsevier Ltd. All rights reserved.

In order to develop models regarding subcooled boiling flow and its heat transfer, we conducted subcooled boiling experiments to investigate subcooled bubble formation and measured development process of the bubbles using a test loop with a vertical annulus flow path under atmospheric conditions. Two-phase flow parameters regarding subcooled boiling, such as instantaneous bubble velocity, turbulent velocity components, void fraction, onset of nucleate boiling (ONB) and onset of significant void (OSV), were obtained with two high-speed video cameras in high resolution in temporally and spatially. The identical bubbles were identified from the two video images taken by the high-speed video cameras, and the trajectories were reconstructed in three-dimensional. The instantaneous bubble velocities and turbulence velocity components regarding bubble transport were quantified with the trajectory data. It was observed that the bubbles were rising along the surface of the heater rod moved largely in the circumferential direction, and it was found from the result that velocity component for circumferential direction was enough large compared with that for radial direction.

Several model parameters affecting void fraction distribution of the subchannel analysis code CTF (previously called COBRA-TF) were selected, and a global sensitivity analysis was performed for the CRIEPI 5 × 5 fuel bundle void tests. Using the sensitivity analysis results as training data, a metamodel was developed with the Kriging method. The response variables such as bundle-averaged void fraction difference and residual void fraction deviation to the prior distributions of the model parameters were estimated from this metamodel, and the posterior distributions of the parameters were obtained so that the response variables would be close to the target distributions. It was confirmed that the bundle-averaged void fraction difference and residual void fraction deviation calculated by CTF could be improved with the parameter set optimized by the data assimilation. Furthermore, using the data assimilation method, the flow characteristics of the cross-flow between subchannels corresponding to the space between fuel rods were back-calculated to reproduce the measured void fraction.

We have developed an efficient, stable device for generating single-micrometer-scale (1-2 mu m) droplets based on the fragmentation of droplet tails by tailing. The device, created by a simple soft lithography process, induced continuous droplet fragmentation by branched channels under low-flow conditions (1-10 mu L/min). The flow rate and surfactant concentration were also important factors for droplet fragmentation. Examining 10 combinations of flow rate and surfactant concentration revealed the optimal conditions, which produced droplets less than 2 mu m in size at a generation rate of 61.1%. With this method, we efficiently encapsulated cell-mimicking microbeads into passively generated single-micrometer-scale microdroplets.

To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant (NPP) Unit-3 (from hereinafter, Unit-3), the meshless, Lagrangian Moving Particle Semi-implicit (MPS) method has been developed for the evaluation of debris-structure interactions in the pedestal region of Unit-3. The developed MPS method incorporated approaches to enhance numerical accuracy and stability, such as second-order corrective matrix and particle shifting technique, and the calculation cost reduction and efficiency improvement strategies to enable plant-scale simulations. A 1/10 scale numerical model of the pedestal region and structures within of Unit-3 and sensitivity analysis conditions were established for simulation cases. The simulation results showed that the convective vapor cooling from the debris surface in the pedestal region and the debris relocation amount/the time intervals between the relocations play important roles in the structure damage due to thermal attacks from the debris.

A subchannel void sensor (SCVS) acquires the two-phase flow in a rod bundle as the time-series data of cross-sectional distributions. Herein, the temperature and pressure ranges of an SCVS were extended to include the rated conditions of boiling water reactors. The improved SCVSs were installed in a 5 × 5 heated rod bundle at eight height levels. In a boiling experiment using the rod bundle, the three-dimensional distributions of the boiling two-phase flow were measured over a wide pressure range (up to 7.2 MPa). The new experimental data were compared with existing experimental data and the results of a subchannel analysis. Experimental results were consistent with those of a high-energy X-ray computed tomography study of a heated rod bundle with the same geometry and under the same heat and flow conditions as those used in our study. The subchannel analysis code reproduced the experimental results fairly well, and the obtained database is applicable for validating and improving thermal-hydraulic analysis codes.

We developed a method for passively controlling microdroplet rotation, including interior rotation, using a parallel flow comprising silicone and sesame oils. This device has a simple 2D structure with a straight channel and T-junctions fabricated from polydimethylsiloxane. A microdroplet that forms upstream moves into the sesame oil. Then, the largest flow velocity at the interface of the two oil layers applies a rotational force to the microdroplet. A microdroplet in the lower oil rotates clockwise while that in the upper oil rotates anti-clockwise. The rotational direction was controlled by a simple combination of sesame and silicone oils. Droplet interior flow was visualized by tracking microbeads inside the microdroplets. This study will contribute to the efficient creation of chiral molecules for pharmaceutical and materials development by controlling rotational direction and speed.

We fabricated a microfluidic device with a three-dimensional (3D) structure and verified its droplet-generation performance for the stable production of droplets of around 10 μm in size. We compared the performance of the 3D device with that of conventional simple T-junction and cross-junction structures. The continuous phase sheared the dispersed phase into droplets from eight directions in the 3D device, compared with only one direction in the T-junction device and two in the cross-junction device. Droplets were produced efficiently over a wide range of fluid properties and flow conditions with the 3D device, unlike with the two conventional planar devices. Fluidic experiments were conducted using mineral oil with a surfactant as the continuous phase, deionized (DI) water as the dispersed phase, and DI water with glycerin to change the viscosity of the dispersed phase. The minimum droplet length was 47.2 μm in the T-junction device, 39.0 μm in the cross-junction device, and 22.4 μm in the 3D device when using a water and glycerin mixture with a viscosity of 9.0 mPa·s. Compared with the conventional devices, smaller droplets were produced using our 3D device, indicating that it has excellent droplet-generation performance.

In order to obtain high-resolution data for modelling of boiling two-phase flow and its validation, we designed and constructed a test loop with a vertical annulus flow path and conducted subcooled boiling experiments to investigate subcooled bubble incipience and its development process under atmospheric condition. Three kinds of the state-of-the art measurement techniques were applied to quantify key parameters such as radial and vertical distributions of void fraction, bubble velocity, interfacial area concentration (IAC), Sauter mean diameters, high-resolution temperature distribution on rod surface, bubble transport behavior, and turbulent velocity components as well as onset of nucleate boiling (ONB), and onset of significant void (OSV).

Natural circulation is a key technology for developing the molten core cooling system without an external power source from the lessons of the severe accident at Fukushima-Daiichi Nuclear Power Station. This study is devoted to quantify the void fraction which is an important parameter for the driving force of natural circulation flow, and to evaluate the effect of the void fraction correlation on the prediction accuracy of the natural circulation flow rate. Test was conducted at atmospheric pressure and room temperature, using the upward air–water two phase flow. Vertical tubes with an inner diameter of 36 and 25 mm were used as the test section. The void fraction was measured by three different methods: quick-closing valve method, pressure drop method, and conductive void-probe method. The following conclusions are obtained from this study: (1) The data of the natural circulation flow rate, void fraction and pressure drop for the upward air–water two phase flow at atmospheric pressure and room temperature were obtained to develop and verify the new model. (2) By improving the void correlation, it was found that the prediction accuracy of the natural circulation flow rate could be improved by about 10% to 5%, that is, the prediction error can be halved in the range of this study. (3) The natural circulation flow rate for 25 mm test section was saturated with increasing the air flow rate at higher air flow condition. The model cannot predict this tendency. From the point of design of the actual molten core cooling system, the model improvements in this region are necessary in the future.

In recent years, research on the application of microdroplets in the fields of biotechnology and chemistry has made remarkable progress, but the technology for the stable generation of single-micrometer-scale microdroplets has not yet been established. In this paper, we developed an efficient and stable single-micrometer-scale droplet generation device based on the fragmentation of droplet tails, called "tail thread mode", that appears under moderate flow conditions. This method can efficiently encapsulate microbeads that mimic cells and chemical products in passively generated single-micrometer-scale microdroplets. The device has a simple 2D structure; a T-junction is used for droplet generation; and in the downstream, multi-branch channels are designed for droplet deformation into the tail. Several 1-2 mu m droplets were successfully produced by the tail's fragmentation; this continuous splitting was induced by the branch channels. We examined a wide range of experimental conditions and found the optimal flow rate condition can be reduced to one-tenth compared to the conventional tip-streaming method. A mold was fabricated by simple soft lithography, and a polydimethylsiloxane (PDMS) device was fabricated using the mold. Based on the 15 patterns of experimental conditions and the results, the key factors for the generation of microdroplets in this device were examined. In the most efficient condition, 61.1% of the total droplets generated were smaller than 2 mu m.

We have established an experimental system to visualize a droplet flow in a simulated BWR fuel sub-channel optically and measure the diameter and velocity of droplet after passing through a spacer. For representative spacers of ferrule and grid type, an effect of them on downstream droplets was evaluated with the experimental system. When a ferrule type spacer was simulated and implemented in both the center and side sub-channel, the vertical velocity of droplets got faster especially in the range of small diameter compared to the case of no spacer. When a grid type spacer was simulated and implemented in the center sub-channel especially, a large dispersion of vertical velocity of droplets occurred especially in the range of small diameter compared to the case of no spacer. By a computational fluid dynamics analysis for gas phase flow to drive the droplets in the sub-channel, it was confirmed qualitatively that the characteristics of droplet behavior observed in this experiment were dependent on the structure and geometry of spacer and sub-channel. Furthermore, it was revealed that a relation between a droplet diameter and velocity can be organized with a non-dimensional function derived from a momentum equation of particle in driving fluid and its drag coefficient has linear correlation with a gas Froude number.

Sea water shall be injected into water-cooled nuclear reactors during severe accidents, which are located along coastal side to flood the nuclear fuel, which is heated by residual heat. Precipitation growth to narrower flow path area is a key to gain the confidence of accident mitigation procedure to cool down the reactor core during accidental conditions. A pool boiling experiment was conducted with a simulated 5 × 5 full-height BWR fuel-rod bundle with condensed (two and half times higher concentration) sea water. The temperature on the center rod surface in the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout below the top spacer escalated temperatures of the heater surface. On the other hand, the heater above the top spacer was cooled stably by pool boiling. An example calculation estimates that the dryout due to salt precipitation may occur 19 h after sea water injection for an ABWR, which had operated at 3.926 GWt for 13 months on the basis of critical dryout concentration of 50 wt%.

The void fraction distribution of a fuel rod bundle in a boiling water reactor is a critical parameter for accurately predicting the optimal thermal margin in the design of a reactor core. The rod bundle configuration, such as a part-length rod (PLR) and water rod, can affect void distribution. To clarify the influence of PLR on void fraction distribution, a boiling flow experiment was conducted using a 5 x 5 heated rod bundle that partially simulated a boiling water reactor (BWR) rod bundle, and three PLRs were arranged in the corner. The cross-sectional void fraction distribution was acquired using high-energy X-ray computed tomography at six height levels for wide flow conditions, system pressures of 0.1 - 7.2 MPa, inlet subcoolings of 20 - 90 kJ/kg, mass fluxes of 500 - 1250 kg/m(2)/s, and linear heat generation rates (LHGR) of 3.2 - 8.6 kW/m. In the PLR region, the local void fraction temporarily decreases because the PLRs disappear, and the flow channel rapidly expands. Together with the downstream PLRs, the voids propagate to the PLR region and concentrate in the center. The void fraction in the corner of the PLR region remains lower. A maximum 26% decrease in the subchannel void fraction was observed in the corner of the PLR region at the system pressure of 7.2 MPa, mass flux of 1.25 x 10(3) kg/m(2)/s, inlet subcooling of 50 kJ/kg, and LHGR of 8.6 kW/m.

To deepen our understanding for the current debris status and investigate the debris-structure interactions in the pedestal region of Fukushima Unit-3, the Moving Particle Semi-implicit (MPS) method is further developed for simulation of multicomponent liquid/solid relocation with solid-liquid phase changes aiming for plant-scale practices. The improvement of the existing MPS method mainly consists of two parts, namely 1) the improvement of numerical stability and accuracy, including a) applying second-order corrective matrix to the particle interaction models and b) particle shifting that can optimize particle configurations; and 2) improvement of calculation efficiency, including a) hybrid OpenMP and MPI parallelization and b) particle type-dependent speed-up algorithms to reduce calculation costs for particles with extremely high viscosity and low velocity. In the current study, the improved MPS method is validated against the experiments carried out at Waseda University, in which molten salt droplets were released to interact with aluminum pillars and solidify on them. Good agreement of the total height of the solidified salt and its distribution on the pillars has been achieved. The successful validation has shown the capability of the current MPS method for simulations of Unit-3 pedestal region.

Steam explosion experiments with a melt layer spreading at the bottom of a shallow water pool, namely the PULiMS-E6 and SES-S1 by KTH, Sweden, were simulated by the steam explosion simulation code, JASMINE. The observed impulses in the experiments were successfully reproduced by simulations with assumed premixing conditions. With those simulation results, the adequacy of the kinetic energy evaluation method used for the experiments were examined by comparison of the kinetic energy directly obtained in the simulation, Ek, and the one evaluated based on the impulse and the water mass limited to the center area above the premixing zone, Ekic. It showed that the impulse based kinetic energy evaluation gives about five times overestimation. The impact of the water pool geometry on the validity of the impulse based kinetic energy evaluation method was further examined by a parametric study with variations of the pool geometry in the simulations of PULiMS-E6 and SES-S1 as well as high pressure bubble expansion simulations. The results for the relation of Ekic/Ek and the geometric factors were consistent between the cases for the experiments and the bubble expansion. The results showed that: (1) for the shallow water pool regime, Ekic/Ek shows a trend of convergence to 4–5, (2) for deep water pool regime, the impulse based kinetic energy evaluation with the whole water mass, Eki, rather than Ekic, gives a good estimation. A set of empirical formulas was obtained for Ekic/Ek.

<p> 流れ現象は日常生活から工業プラントまで身近に存在するが，それを定量的なものとして把握するために可視化技術が発展してきた。原子炉燃料集合体はその複雑な構造内部を水が流れながら除熱や中性子減速材の役割を果たす。本稿では，当所が研究開発を進めている燃料集合体内の沸騰二相流計測技術に焦点をあてる。これまでに開発した計測技術の測定原理や計測機器の概略を示し，模擬燃料集合体を用いた計測事例を紹介する。</p>

A boiling experiment was conducted to acquire a threedimensional void fraction distribution in a rod bundle with part length rods. The test section was a 5 × 5 heated rod bundle that partially simulated a BWR rod bundle, and three PLRs were arranged together in the corner. The heated length of the fulllength rod was 3.71 m, which is equal in length to an actual fuel rod in BWRs, whereas the heated length of the PLR was 1.85 m. The rod bundle exhibited an axially and radially uniform heat flux. Further, the local void fraction was quantified by normalizing the intensities in the CT images of the boiling twophase flow with those obtained under the liquid-phase and gasphase conditions. The cross-sectional void fraction distribution was acquired at six height levels. The experimental results exhibited the void distribution around PLRs with respect to a wide range of flow conditions, inlet temperatures, inlet flow rates, bundle thermal powers, and system pressures.

Gas-liquid two-phase flow in a stagnant pool is an important phenomenon in designing and operating industrial facilities. When gas is mixed or boiling occurs in stagnant water, the actual water level appears higher than the original water level. The actual water level is called a two-phase mixture level and largely depends on the flow channel geometries, dimensions, and flow conditions. This study focuses on the influence of channel geometries, circular pipes and rod bundles, on the two-phase mixture level and its fluctuation behavior. An air-water experiment using circular pipes with inner diameters of 50 and 224 mm and 5 x 5 and 10 x 10 rod bundles was conducted, and the two-phase mixture level swell was visually observed. As the inlet gas flow rate increased, the two-phase mixture level basically increased regardless of the channel geometry. The fluctuation amplitude was remarkably increased by formulating the slug bubbles covering the entire diameter in the small pipe with a diameter of up to 50 mm. In the rod bundles and large pipe with a diameter of 224 mm, no slug bubble was sustained, and the two-phase water level and its fluctuation amplitude were relatively small compared with those of the small pipe.

Carbon-coated SiO (SiO-C), which is a high-capacity anode material, experiences a significant capacity drop in the initial charge-discharge cycles. In contrast, Li-doped SiO-C (Li-SiO-C), which has been recently developed, exhibits a significantly smaller capacity drop. To explain this difference, we performed a detailed investigation of the structures and phase changes associated with the charge-discharge cycling of these materials by comparing their Si structures and electronic states obtained from solid-state magic-angle spinning nuclear magnetic resonance and Si K-edge X-ray absorption fine structure measurements. The results show that, in the case of SiO-C, the Li(4)SiO(4)generated during charge is partially decomposed during discharge in the initial charge-discharge cycles. These generation and decomposition behaviors are most intense during the first 20 cycles. We believe that this phenomenon is the cause of the increased irreversible capacity observed in the initial cycles of SiO-C. In addition, we confirmed that Li2SiO3, a component of Li-SiO-C, is relatively stable electrochemically, although some of it gradually converts into Li(4)SiO(4)during charge-discharge cycling. The presence of Li(2)SiO(3)at the outset implies that less Li(4)SiO(4)is generated during charging compared to SiO-C, which we believe explains the lack of a significant capacity drop in the initial cycles of Li-SiO-C.

An analytical model was developed to describe thermal stratification in a primary containment vessel (PCV) and transient thermal-hydraulics coupled with a reactor pressure vessel (RPV) using TRACE code version 5.0 patch level 4. Geometries of a dry well (D/W) and a suppression chamber (S/C) were represented by a nodalization of TRACE code to simulate multi-dimensional flow in the PCV. An additive loss coefficient (so called ‘K-factor’) was focused as a sensitivity parameter to limit flow rate in a pool. For the first step, a validation analysis was conducted against a steam discharge experiment of S/C. The TRACE result was in good agreement with the measurement and showed a thermally-stratified temperature distribution in the S/C pool. For the second step, an analysis to simulate the accident at Fukushima Daiichi Unit 3 power plant (1F3) was conducted. It was proved to be able to explain the pressure increase in the PCV at the beginning of accident by demonstrating thermal stratification in the S/C pool. Sensitivity study revealed an optimal K-factor value for a macroscopic viscous drag in a liquid phase fluid to demonstrate thermal stratification in a pool.

Liquid ZrO2 is one of the most important materials involved in severe accident analysis of a light-water reactor. Despite its importance, the physical properties of liquid ZrO2 are scarcely reported. In particular, there are no experimental reports on the viscosity of liquid ZrO2. This is mainly due to the technical difficulties involved in the measurement of thermo-physical properties of liquid ZrO2, which has an extremely high melting point. To address this problem, an aerodynamic levitation technique was used in this study. The density of liquid ZrO2 was calculated from its mass and volume, estimated based on the recorded image of the sample. The viscosity was measured by a droplet oscillation technique. The density and viscosity of liquid ZrO2 at temperatures ranging from 2753 K to 3273 K, and 3170 K–3471 K, respectively, were successfully evaluated. The density of liquid ZrO2 was found to be 4.7 g/cm 3 at its melting point of 2988 K and decreased linearly with increasing temperature, and the viscosity of liquid ZrO2 was 13 mPa at its melting point.

<p>Filtered containment venting systems (FCVS) installed in the exhaust systems reduce the release of the radioactive materials by the multistage filters. Venturi-scrubber, wet scrubbing, static mixer and metal-fiber filter remove aerosol. This study has focused on the venturi-scrubber which remove aerosol by collision of droplets with particles and conducted evaluation of aerosol removal performance and hydro-dynamic behavior. Performance of venturi scrubber depend on its geometries such as throat cross sectional area, spread angle of expanding section. This study used several type of venturi scrubbers, the test results show that the aerosol DF depend on aerosol diameter and the venturi-scrubber used in this study shows DF ≥ 100 for aerosol (diameter: 1.0 μm). Moreover, this study evaluated the aerosol removal performance using CFD calculation and previous model (Ueoka and Kawakami model). By adjusting the constant included in the collision coefficient (ε), the venturi scrubber performances predicted from the CFD calculation are in good agreement with the experimental results.</p>

In order to gain the reliability of subchannel analysis on three-fluid two-phase flow in nuclear fuel assemblies, implemented models are expected to describe the detailed three-dimensional two-phase flow in fuel assemblies. Especially, crossflow model limits the prediction performance of subchannel analysis codes, it is important to develop a model that can analyze the phenomenon appropriately. In this study, CTF results are validated against the NUPEC BWR Full-Size Fine-mesh Bundle Test (BFBT). In BFBT, void fraction distribution across 8×8 rod bundles was measured to confirm the effects of radial/axial power distribution and unheated rods. Moreover, uncertainties of void fraction were quantified. In order to evaluate the prediction performance of the CTF code for the BFBT, bundle-averaged void fraction difference and the residual void fraction difference for each subchannel were defined. Subchannels were classified into several groups considering the grid spacer pressure loss coefficients set for each channel and the characteristics of the subchannel considering location, such as corner subchannel, adjacent corner subchannel, etc. Sensitivity parameters affecting void fraction and/or cross flow were selected with the Kriging method, and the response surface model represented by these sensitivity parameters was created. the simulation-driven MCDA method using the alternative model was applied for optimizing sensitivity parameters by data assimilation, and a set of parameters to accurately calculate the bundle-averaged void fraction difference was identified with the Metropolis method. CTF analysis with the parameter set identified by the data assimilation was conducted, and it was confirmed that the average value of the bundle-averaged void fraction improved with the parameter set by the data assimilation so that the predicted value would match the experimental value.

In the case of a loss in a coolant accident when the water level of the reactor core falls below the top of active fuel (TAF), the actual water level, namely the two-phase mixture level, is an important factor to predict the coolability of the core. The two-phase mixture level depends on the liquid inventory and temporal and spatial variations of the void fraction in the core, and can be greatly affected by the system pressure. A boil-off experiment in which the water level decreased by evaporation without any water supply, was conducted using a 5×5 heated rod bundle over a wide pressure range from 0.1 to 7.2 MPa. The heated length of the rod was 3.71 m, which is equal in length to an actual fuel rod in boiling water reactors (BWRs). The 5×5 rod bundles had an axially and radially uniform power profile, and eight pairs of fine thermocouples were embedded in the heated region to acquire the axial distribution of the rod surface temperature. The temporal fluctuation of the two-phase mixture level was visualized by a linear-accelerated driven X-ray real-time radiography at 400 frames per second. The experimental results exhibit the influence of the two-phase mixture level variation on the rise of rod-temperature under boil-off conditions.

In a severe accident of a light water reactor, the reactor pressure vessel (RPV) lower head may fail due to ablation at the vessel wall boundary involving natural convection of molten core materials. Accurate prediction of RPV lower head failure is essential for assessing severe accident progression and improving accident management because it greatly influences the subsequent ex-vessel accident progressions. However, there have been still large uncertainties about RPV lower head failure mode in the Fukushima Daiichi Nuclear Accident in 2011. The Lagrangian based MPS (moving particle semi-implicit) method has advantage of analyzing such phenomena involving complex interfaces and liquid-solid phase changes over other Eulerian mesh-based method. In the preceding study, small-scale Pb–Bi hemisphere vessel ablation experiment, with silicone oil as simulated molten core, was reproduced qualitatively by original MPS method. However, ablation mechanism associated with natural convection of the high temperature liquid could not be discussed because of significant influence of numerical discretizing error. In this study, the improved MPS method coupling corrective matrix in the particle interaction model which largely suppress the numerical fluctuation was adopted to analyze the experiment. The results show that the ablated metal relocation may enhance convective heat transfer in the downstream. As a result, ablation of the vessel wall extends from the level, close to the silicone oil surface down to the bottom of the vessel rather than previously simulated localized ablation near the silicone oil surface.

Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped.This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5 x5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40 degrees C, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.

The bubble pairing scheme was devised to quantify three-dimensional velocity of each bubble. We used two sets of Wire-Mesh Sensors to identify locations of each bubble according to bubble identification algorithm, which was developed by HZDR. The devised scheme was applied to the vertical upward air-water flow at 0.64 m/s for both air and water superficial velocities in a large diameter pipe (i.d. 224 mm). The bubble pairing scheme visualized the developing process of two-phase flow: large bubbles coalesced with each other to move toward the center, while the rest of bubbles broke up into smaller bubbles and decelerated.

In a severe accident of a light water reactor, the corium spreading behavior on a containment floor is important as it may threaten the containment vessel integrity. The Moving Particle Semi-implicit (MPS) method is one of the Lagrangian particle methods for simulation of incompressible flow. In this study, the MPS method is further developed to simulate corium spreading involving not only flow, but also heat transfer, phase change and thermo-physical property change of corium. A new crust formation model was developed, in which, immobilization of crust was modeled by stopping the particle movement when its solid fraction is above the threshold and is in contact with the substrate or any other immobilized particles. The VULCANO VE-U7 corium spreading experiment was analyzed by the developed MPS spreading analysis code to investigate influences of different particle sizes, the corium viscosity changes, and the “immobilization solid fraction” of the crust formation model on the spreading and its termination. Viscosity change of the corium was influential to the overall progression of the spreading leading edge, whereas termination of the spreading was primarily determined by the immobilization of the leading edge (i.e., crust formation). The progression of the leading edge and termination of the spreading were well predicted, but the simulation overestimated the substrate temperature. Further investigations may be necessary for the future study to see if thermal resistance at the corium-substrate boundary has significant influence on the overall spreading behavior and its termination.

A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of void fraction in a 5x5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied to a boiling two-phase flow experiment under the atmospheric pressure condition assuming at an accident or in a spent fuel pool in a boiling water reactor (BWR). The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. The wire electrodes of 0.2 mm in diameter are arranged in lattice patterns between the rod bundle, while the electric conductance value in a region near one wire and another corresponds to local void fraction in the central-subchannel region. The local void fractions at 32 points (= 6x6-4) can be obtained as a cross-sectional distribution. The local void fractions near the rod surface at 100 points (= 4x25) can be also estimated by the conductance value in a region between one wire and one rod. The devised sensors are installed at five height levels along the axis to acquire two-phase flow behavior. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The temporal resolution of void fraction measurement is 1600 frames (cross-sections) per second. The axial and radial power profile of the heated rod bundle are uniform, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various inlet flow velocity, rod bundle power and inlet subcooling conditions. The experimental results were presented by the axial and cross-sectional distributions of void fraction, phasic velocity and bubble-chord length.

The interfacial behavior of upward liquid film flow is an important phenomenon to evaluate interfacial transfer accompanying the entrainment and deposition of droplets. This research focuses on a vertical annular channel, and an air-water liquid film flow experiment was conducted under atmospheric pressure conditions. The diameters of inner and outer pipes in the annular channel were 12 and 18. mm respectively. The experiment featured multi-point electrode sensors installed in both the inner and outer pipe surface at the same height, and the ability to measure the liquid film distribution on both surfaces in the annular channel simultaneously. As for the sensor structure, 10. ×. 32 measuring points were arranged in a lattice pattern on the sensor surface and the spatial resolution was 2. ×. 2. mm, hence the liquid film thickness distribution could be measured rapidly, at over 1250 slices per second. Since the sensor was manufactured by a flexible multilayer substrate, it was applicable to a cylindrical channel surface. In the experiment, water was supplied from the inner pipe surface and uniformly distributed in the circumferential direction, whereupon liquid film distributions were measured 300. mm downstream from the water supply position. The time series data of the liquid film distribution demonstrated circumferential distributions of liquid film thickness and interfacial wave velocity. When the superficial gas velocity was smaller than 20. m/s, a liquid film formed on both inner and outside pipe surfaces, regardless of the superficial liquid velocity. With increasing superficial gas velocity, the film thickness of the outer pipe surface became thinner than that of the inner pipe surface. Measurement of the liquid film thickness on both surfaces of the annular channel also showed that a liquid slug with wavelength of several millimeters passed concurrently through both surfaces in the annular channel.

When there is no power for cooling the spent fuel pool and conditioning the air in a boiling water reactor (BWR) reactor building, water vapor is generated from the pool and it affects the atmosphere in the building. To consider the impact of the steam in preparing emergency operation procedures, the building atmosphere under various conditions is to be evaluated with reasonably low computational cost. A lumped parameter model to predict the transient behavior of the building atmosphere was developed, in which the evaporation from the spent fuel pool and the condensation to the wall were taken into consideration. A transient behavior of temperature and vapor concentration in a BWR operating floor was predicted with the model. The results and the prediction speed were compared to those of a three-dimensional computational fluid dynamic calculation, and it was confirmed that the model could obtain almost the same results about 280,000 times faster. Parameter studies are conducted with the model, and dominant parameters to the evaporation and the condensation were clarified.

In order to evaluate residual the stress distribution in oxides formed on zirconium alloys, synchrotron X-ray diffraction (XRD) was performed on the oxides formed on Zircaloy-2 after autoclave treatment at a temperature of 360° C in pure water. The use of a micro-beam XRD and a micro-sized cross-sectional sample achieved the detailed local characterization of the oxides. The oxide microstructure was observed by TEM following the micro-beam XRD measurements. The residual compressive stress increased in the vicinity of the oxide/metal interface of the pre-transition oxide. Highly oriented columnar grains of a monoclinic phase were observed in that region. Furthermore, at the interface of the post-first transition oxide, there was only a small increase in the residual compressive stress and the columnar grains had a more random orientation. The volume fraction of the tetragonal phase increased with the residual compressive stress. The results are discussed in terms of the formation and transition of the protective oxide.

The eutectic reaction phenomenon was analyzed based on the Moving Particle Semi-implicit (MPS) method. The improved MPS code was applied to three-dimensional Pb-Sn system experiments conducted at Central Research Institute of Electric Power Industry (CRIEPI). The experiments dealt with solid-solid contact materials at 225 and 205 °C. The mass diffusion process was modeled based on Fick's second law. The criterion of eutectic reaction was modeled based on binary phase diagram. The calculated penetration rates, i.e. height reduction rates, were compared with the experimental measurements. The results obtained by the MPS simulations exhibit good agreement with the experiments carried out at 205 and 225 °C. MPS simulation predicted that Sn particle will liquefy earlier than Pb particle.

Three-fluid dam-break experiments were conducted to observe the mixing and stratification processes of three immiscible fluids (two-liquids and one gas) by gravity. Two liquids (silicone oil and salt water) were separated with a vertical wall and filled in a rectangle container. Fluid motions are visualized by four sets of video cameras which are synchronized with each other. Parametric study reviles the effects of two key fluid properties: kinematic viscosity (or molecular weight) for silicone oil and density (or concentration) for sodium chloride aqueous water. After the withdrawal of a vertical partition plate, two liquids intersected earlier at the center, while the fluids stuck on the walls. The kinematic viscosity and density difference affect the three-dimensional mixing and stratification processes significantly. These visual databases are suitable for a code validation on the interfacial phenomena. Computational multi-phase fluid dynamics analysis was conducted with Star CCM+ version 8.04. The numerical results agree with experimental results accordingly for all those key parameters. The physics behind the contact angle and interfacial tensions are thoroughly investigated parametrically.

Purpose: In the present study, we compared 12- with 16-core biopsy in patients with prostate-specific antigen (PSA) levels of 4.0-20.0 ng/mL.Materials and Methods: Between 2003 and 2010, 332 patients whose serum PSA level was between 4.0 and 20.0 ng/mL underwent initial transrectal ultrasound (TRUS)-guided needle biopsy. Of those patients, 195 underwent 12-core biopsy and 137 underwent 16-core biopsy.Results: In the 12-core prostate biopsy group, 66(33.8%) patients were found to have prostate cancer. On the other hand, in the 16-core prostate biopsy group of 137 patients, 61(44.5%) were found to have prostate cancer. Among all patients, the prostate cancer detection rate was slightly higher in the 16-core biopsy group than in the 12-core biopsy group. Moreover, in patients with prostate volume > 30 mL or PSA density (PSAD) <0.2, the prostate cancer detection rate was significantly higher in the 16-core biopsy group than in the 12-core biopsy group. There was no significant difference in pathological tumor grade, indolent cancer probability, or biopsy complication rate between the two groups.Conclusion: In order to detect prostate cancer, 16-core prostate biopsy is safe and feasible for Japanese patients with serum PSA level of 4.0-20.0 ng/mL.

A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of a void fraction in a 5x5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied in a boiling two-phase flow experiment under the atmospheric conditions assumed in an accident and spent fuel pool. The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. Wire electrodes 0.2 mm in diameter are arranged in latticed patterns between the rod bundle, while a conductance value in a region near one wire and another gives a local void fraction in the central-subchannel region. 32 points (= 6x6-4) of the local void fraction can be obtained as a cross-sectional distribution. In addition, a local void fraction near the rod surface can be estimated by a conductance value in a region near one wire and one rod using the simulated fuel rods as rod electrodes, which allows 100 additional points (=4x25) of the local void fraction to be acquired. The devised sensors are installed at five height levels to acquire two-phase flow dynamics in an axial direction. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The time resolution of void measurement exceeds 800 frames (cross-sections) per second. The heated rod bundle has an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various experimental conditions, such as inlet-flow velocity, rod-bundle power and inlet subcooling. The experimental results exhibited axial and radial distribution of two-phase flow structures, i.e. void-fraction and phasic-velocity distributions quantitatively.

In order to gain the best use of filtered containment venting systems (FCVSs), the decomtamination factor of FCVSs is to be investigated as a function of system parameter including steam flow rate, pressure, temperature, water level, and operating time. A full-height test facilities were designed and constructed in Central Research Institute of Electric Power Industry (CRIEPI), Japan to evaluate the decontamination factor (DF) in FCVSs. The target types are the orifice and the venturi FCVSs. The height and the internal diameter of the cylindrical test vessel is 8 m and 0.5 m. Bubbly flows were visualized through the view window up to 0.8 MPa and 170 °C. Steam bubbles in 0.2 wt% sodium thiosulfate and 0.5 wt% sodium hydroxide were found to be much smaller than those in water. The DF were evaluated for the aerosol, elemental iodine and organic iodine. The installed aerosol optical spectrometer measures the number density and the diameter of aerosols. The concentrations of elemental iodine were quantified with an inductively-coupled plasma with mass spectrometry (ICP-MS). The concentration of organic iodine was quantified with a gas chromatography with mass spectrometry (GC-MS). In order to investigate two-phase flow dynamics in the vessel, separate effect tests were conducted with air-water test facility. The height of cylindrical test vessel is 8 m. Visual observation was conducted for two internal diameter levels: 0.05 and 0.5 m. High speed video frames were recorded through the transparent (acrylic) vessel wall. Wire-Mesh Sensors (WMS) were installed to acquire a cross-sectional void fraction to compare with DF in the facility. On the basis of the obtained database, we develop the FCVSs performance evaluation technique and propose an optimal FCVSs operation method for a further safety improvements of the nuclear power plant.

Given the severe accident of a light water reactor (LWR), stratification and solidification/melting are important phenomena in melt corium behavior within the reactor lower head, influencing the decay heat distribution and ablation of penetration tube and vessel wall. Numerical calculation is a necessary and effective approach for mechanistic study of local melt corium behavior. In this study, the improved moving particle semi-implicit (MPS) method was applied for investigating the stratification and solidification/melting phenomena. The implicit viscous term calculation technique and stability improvement technique were adopted to enable MPS to simulate the stratification process of materials with high viscosity in phase transition stage. The solid-liquid phase transition model was also coupled with MPS method. The validation experiment was carried out with low-melting-point metal tin and NeoSK-SALT. The layer configurations and temperature profiles obtained from MPS calculation showed good agreement with the experimental results. Meanwhile, the calculation results indicated that the material freezing behavior could affect the layer formation, and the layer configurations also significantly influenced the temperature profiles and heat flux distributions. The present results demonstrated that MPS method has the capacity to understand the local melt behavior in detail that is relevant to stratification and phase transition. © 2014 Elsevier B.V.

Hydrogen diffusion behavior in a cesium adsorption vessel is assessed. The vessel is used to remove radioactive substance from contaminated water, which is proceeded from Fukushima accident. Experiment and numerical calculation are conducted to clarify the characteristics of natural circulation in the vessel. The natural circulation arising from the temperature difference between inside and outside the vessel is confirmed. We develop an evaluation model to predict the natural circulation and its prediction agrees well with the results obtained by the experiment and the calculation. Using the model, we predict steady and transient behavior of hydrogen concentration. Results indicate that hydrogen concentration is kept lower than the flammability limit when the short vent pipe is open. © 2014 Atomic Energy Society of Japan. All rights reserved.

In order to conform to the new regulatory standard in Japan, seawater is regarded as the alternative water source both for BWRs (Boiling water reactors) and PWRs (Pressurized water reactors). For preventing further accident evolutions occurred in Fukushima Daiichi nuclear power plants, seawater was injected into the reactors for more than one week. With long-term seawater injection, sea salt compositions are condensed and many of them will precipitate when saturated concentrations are exceeded. Unlike corrosion issues, impacts of sea salt precipitation on the heat removal has not been studied widely in the past because it has not been regarded as the alternative water source before Fukushima Daiichi nuclear power plants accident. The existent knowledge base of boric acid precipitation under LOCA conditions was studied. Based on the existent study on impacts of boric acid precipitation under LOCA conditions, the experimental project of seawater consisting of four experiments was proposed. The existent database of physical properties, such as viscosity, of seawater is rather poor under severe accident conditions. In addition, it is likely that boric acid will be injected with water or seawater to prevent re-criticality. It is known that physical properties of boric acid vary widely under high temperature conditions. Measurement of viscosity of the seawater-boric acid mixture was conducted in high temperature using a rotational viscometer. Under conditions equivalent to the estimated bulk coolant conditions under a long-term cooling phase of severe accidents. In this range, no obvious change of viscosity is expected. Then a detailed structure of seawater precipitates was observed using a mockup fuel bundle with 5x5 in the square lattice and 500 mm in the length. Images of precipitates were taken using the X-ray CT. The water level, the concentration of sea salt and the heat flux are employed as experimental parameters. The heat flux, bubble stirring in downstream of spacers and heat loss by a non-heated channel box were identified as influential factors to local and overall precipitates in fuel bundles.

The carbon-doped copper oxide is devised as an electrocatalyst to reduce carbon dioxide under ambient pressure and temperature. The electrode was immersed into a KHCO3 aqueous solution with CO2 bubbling. The electric potential was maintained at -1.64 V vs SHE. The electrode prepared at 900 oC gives the maximum production rate of ethylene (25 %), ethanol (6.9 %) and 1-propanol (3.6 %). The production rate of methane, from which is harmful to separate ethylene, was suppressed to one fifteenth of that of ethylene. In contrast to a thermally-oxidized copper-oxide layer, the doped-carbon and a high ratio of Cu2O to CuO in the devised electrocatalyst may result in the higher productivity and selectivity. ©2013 The Japan Society of Mechanical Engineers.

Stratification behavior is of great significance in the late in-vessel stage of core melt severe accident of a nuclear reactor. Conventional numerical methods have difficulties in analyzing stratification process accompanying with free surface without depending on empirical correlations. The Moving Particle Semi-implicit (MPS) method, which calculates free surface and multiphase flow without empirical equations, is applicable for analyzing the stratification behavior of fluids. In the present study, the original MPS method was improved to simulate the stratification behavior of two immiscible fluids. The improved MPS method was validated through simulating classical dam break problem. Then, the stratification processes of two fluid columns and injected fluid were investigated through experiments and simulations, using silicone oil and salt water as the simulant materials. The effects of fluid viscosity and density difference on stratification behavior were also sensitively investigated by simulations. Typical fluid configurations at various parametric and geometrical conditions were observed and well predicted by improved MPS method. © 2013 Elsevier B.V. All rights reserved.

A gas-liquid two-phase flow in a large diameter pipe exhibits a three-dimensional flow structure. The wire-mesh sensor (WMS) can acquire a quasi-three-dimensional void fraction distribution. Furthermore, the WMS can acquire a phasic-velocity distribution on the basis of the time lag of void signals between both sets of WMS. Previously, the acquired phasic velocity was one-dimensional distributions.The authors propose a method to estimate the three-dimensional phasic-velocity distribution from the same WMS data. A three dimensional velocity vector was determined on the basis of cross-correlation analysis. The flow direction is determined by the WMS measuring-point combination, whereby the cross-correlation coefficient between both sets of WMS measuring points reveals the peak. In addition, the flow structure can be extracted by size on the basis of a wavelet analysis.The proposed method was applied for two sets of 64. ×. 64 mesh sensors in an air-water flow in a vertical pipe with inner diameter of 224. mm. The proposed method can successfully visualize a swirl flow structure where large and small bubbles tend to move respectively in inward and outward directions in turn. © 2012 Elsevier Ltd.

The TRACE code was validated against the flashing-induced density wave oscillation in the SIRIUS-N facility at low pressure (from 0.1 to 0.5 MPa) as a part of the international CAMP-Program of USNRC. The SIRIUS-N facility is a scaled copy of natural circulation BWR (ESBWR). Stability map of TRACE agrees with that of SIRIUS-N facility at low subcooling region, though instability observed in the lower heat flux and higher subcooling region from the stability limit of experiment. The TRACE code demonstrates the flashing-induced density wave oscillation characteristics: The oscillation period correlates well with the transit time of single-phase liquid in the chimney regardless of the system pressure, inlet subcooling, and heat flux. Unlike Type-I and II density wave oscillations, the inlet or exit throttling does not affect stability boundary and oscillation amplitude of flashing-induced density wave oscillations significantly. Increasing pressure decreases oscillation amplitude. The comprehensive validation confirms that the TRACE code can demonstrate thermal-hydraulic stability of natural circulation BWRs. Copyright © 2012 by ASME.

Cladding materials with superior corrosion resistance and anti-hydrogen pickup have been developed for high burnup nuclear fuel. We have suggested a surface modification of the cladding materials for this purpose and invented a new surface modification method "Fresh-Green". The Fresh-Green treatment oxidizes and carbonizes a material surface in the same process. Zircaloy-2 with the Fresh-Green treatment showed the improvement of corrosion resistance in autoclave tests. In order to investigate the effect of surface modifications on the corrosion resistance, a synchrotron radiation experiment and a TEM observation were performed on different oxide layers formed on Zircaloy-2. The oxide layers were formed by air-oxidation, an autoclave test and the Fresh-Green treatment. Crystal structures of all the samples were transformed as Zr > Zr3O > tetragonal ZrO2 > monoclinic ZrO2 from the matrix to the surface. Columnar grains of monoclinic zirconia were arranged unidirectionally in the Fresh-Green oxide layer treated at a low temperature. Diffusing capacity for oxygen influenced the crystal structure of the oxide layers. © 2011 Elsevier B.V. All rights reserved.

A high-temperature stainless-steel sphere was immersed into various salt solutions to investigate the film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a digital video camera. Both surface temperature of the sphere and solid-liquid contact behavior were measured. Results of the experiment showed that salt additives enhanced condensation heat transfer, and the observed vapor film was thinner. Furthermore, the frequency of direct contact between the sphere surface and coolant increased. The quenching temperature increased with increased salt concentration, and was highly correlated with ion molar concentration, which represents the density of ions regardless of the type of salt. © 2010 Wiley Periodicals, Inc.

We propose CANOPUS method, which is innovative rapid cooling and liquid atomization process utilizing a sustainable small-scale spontaneous vapor explosion with a vapor explosion promoter as a quenchant. In this manner, one can utilize vapor explosion securely and efficiently to produce the fine amorphous powder. The cooling rate of the CANOPUS method is up to 1.5·108 K/s, which is 280 times higher than that of the conventional water atomizing and is thousands times higher than that of the widely-used gas atomizing. CANOPUS method was successfully applied to a highly viscous (18 Pa s) Mora coal gasification slug, resulting in 30 m powders, which can be used for one of the raw materials of cement. CANOPUS process produces the following highly functional powders: (1) homogeneous materials without segregation, (2) amorphous powders, (3) powders with control grain boundary size, and (4) fine-scale powders. The CANOPUS method allows us to produce new functional materials with excellent tenacious, corrosion resistant, and soft magnetic properties for more practical use, which was difficult to amorphize and control grain boundary size by the other cooling methods so far.

This study examines a corrosion control technique for corrosion-resistant materials or of stainless steel in piping for nuclear reactors. This employs an effect of Radiation Induced Surface Activation (RISA). The experimental results revealed: (1) The mechanism behind the corrosion control proposed by the previous report was confirmed to be appropriate. This via tests that measured the amount of dissolved oxygen and iron ions, in the solution. (2) The corrosion control technique was confirmed to be useful for stainless steel with any kind of metal oxide film coating on the surface. (3) It was also shown to be useful even in actual seawater, due to biological effects, which is a far more severe environment for corrosion control than simple salt water. The corrosion control technique for corrosion-resistant material using RISA in seawater has therefore been shown to offer a significant potential for practical applications in naval architecture and marine structures. Copyright © 2010 by ASME.

Integral Effects Test (IET) was conducted to investigate the effects of flow redistribution during the generator load rejection event by using the SIRIUS-F facility, which simulates boiling two-phase flow in a BWR core. Owing to the automatic controllers of a recirculation pump inverter and fine-control valves in the facility, the time series of signals of heat flux and mass flux were observed to agree well with those of target rapid flow-decrease events in the previous experimental series. This paper addresses the simulated generator load rejection event, during which the flow and power gradually decrease and the flow takes a turn toward recovery. As a result of the two-parallel channel experiment, mass flux of a hot channel is lower than that of the other during the initial stage. When the void fraction becomes smaller, mass flux of the hot channel is observed to become higher. This phenomenon can be accurately demonstrated with the TRAC-BF1 code as well. The code does, therefore, predict the boiling two-phase flow in a BWR core even at such flow-decrease event. During the event, differential pressure along each channel between the upper and lower plena decreases by several tens of kPa. The relative perturbations of the differential pressure between both channels, however, remain less than 0.4 %, which is a significantly small amount. In conclusion, the differential pressures between the upper and lower plena of two-parallel channels are, therefore, identical to each other regardless of the power. Copyright ©2009 by ASME.

A high-temperature stainless-steel sphere was immersed into Al 2O3 nanofluid to investigate film boiling heat transfer and collapse of vapor film. Surface temperature is referred to the measured value of thermocouples embedded into and welded onto a surface of the sphere. A direct contact between the immersed sphere and Al2O3 nanofluids is quantified by the acquired electric conductivity. The Al 2O3 nanofluid concentration is varied from 0.024 to 1.3 vol%. A film boiling heat transfer rate of Al2O3 nanofluid is almost the same or slightly lower than that of water. A quenching temperature rises slightly with increased the Al2O3 nanofluid concentrations. In both water and Al2O3 nanofluid, the direct contact signals between the sphere and coolant were not detected before vapor film collapse. Copyright ©2009 by ASME.

A high-temperature stainless steel sphere was immersed into various salt solutions to investigate film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a high-speed digital-video camera. Because the salt additives enhance the condensation heat transfer, the observed vapor film was thinner. The surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature because the frequency of direct contact between the sphere surface and the coolant increased. Quenching temperature increases with increased salt concentration. The quenching temperature, however, approaches a constant value when the salt concentration is close to its saturation concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts. © 2009 by ASME.

Crud deposition on fuel cladding surface would become significant issue in Japanese PWRs, because the large amounts of crud deposition has led to an increase of the dose rate in the primary coolant system and become a root cause of axial offset anomalies (AOA). In order to clarify the contribution factors of the deposition, the effects of heat flux, Ni/Fe ratio and nickel concentration in the test solution on the crud deposition were investigated in a simulated Japanese PWR fuel cycle chemistry at 325°C under sub-cooled boiling and non-irradiated condition. From the test results, it was revealed that the crud layer composed of NiFe2O4 and NiO was formed on the fuel cladding. The amounts of deposited crud layer increased with increase of heat flux and Ni/Fe ratio in the test solution. Ni contents in the crud layer increased with increase of Ni concentration in the test solution.

When a semiconductor film is irradiated by γ-rays, excited electrons are transferred to a base metal in contact with the film, resulting in cathodic-anodic reactions and surface activation of the metal oxide film. The authors first produced radiation-induced surface activation (RISA) in 2000 and have used it in the development of a new corrosion protection method. This report describes a corrosion mitigation technique based on RISA to prevent crevice corrosion in stainless steel, using low-intensity radiation. Experimental results show that an electrode potential of -100 mV vs. Ag/AgCl was produced and maintained on TiO2-coated SUS304 stainless steel specimens immersed in artificial seawater and in close contact with a small, sealed 60Co source (external irradiation) or activated by neutron irradiation to become self-exciting, with no corrosion observed for more than 7 days. In contrast, the potential of a specimen without a radiation source decreased to less than -280 mV vs. Ag/AgCl and crevice corrosion occurred beneath the O-ring within a few days. The corrosion control mechanism was explored by measurement of dissolved oxygen and iron ions in the solution.

When a metal oxide is irradiated by gamma rays, the irradiated surface becomes hydrophilic. This surface phenomenon is called as radiation induced surface activation (RISA). In order to investigate radiation-induced and photo- induced hydrophilicity, the contact angles of water droplets on a titanium dioxide surface were measured in terms of irradiation intensity and time for gamma rays of cobalt-60 and for ultraviolet rays. Reciprocals of the contact angles increased in proportion to irradiation time before the contact angles reached their super-hydrophilicity state. The reciprocals of contact angles correlate well with integrated intensity by a straight line, regardless of the irradiation intensity and time. Radiation induced and photo-induced hydrophilicity phenomena are identical to each other in this regard. In addition, an effect of ambient gas was investigated. In pure argon gas, the contact angle remains the same against the irradiation time. This clearly shows that a certain humidity in ambicnt gas is required to take thc placc of RISA hydrophilicity. A single crystal titanium dioxide (100) surface was analyzed by X-ray photoelectron spectrometry (XPS). After irradiation with gamma rays, a peak was found in the 0 Is spectrum, indicating the adsorption of dissociative water to a surface 5-fold coordinate titanium site, and the formation of a surface hydroxyl group. We conclude that the RISA hydrophilicity is caused by chemisorption of the hydroxyl group on the surface. © 2008 by ASME.

Small-scale experiments have been conducted to investigate the triggering mechanism of vapor explosions. In order to attain good repeatability and visibility, a smooth round water droplet was impinged onto a molten alloy surface. This configuration suppresses premixing events prior to triggering. Six molten metal and alloys were used as the pool liquid. The lower limit of the contact temperature in the vapor explosion region closely agrees with the spontaneous bubble nucleation temperature of water. The upper limit of the initial molten alloy temperature decreases when an oxide layer forms on the surface causing an increase of the emissivity of thermal radiation that has a stabilizing effect on the vapor film. When an oxide layer was formed on the surface, a water droplet was occasionally entrapped into a molten alloy dome, since the oxide layer prevents the droplet from evaporating coherently. The vapor explosion region obtained for the mirror surface is a conservative estimate, since that for the oxide surface fell into the internal region of the vapor explosion for the mirror surface. © 2008 Elsevier Ltd. All rights reserved.

A high-temperature stainless-steel sphere was immersed into various salt solutions to test film boiling behavior at vapor film collapse. The film boiling behavior around the sphere was observed with a high-speed digital-video camera. Because salt additives enhanced condensation heat transfer, the observed vapor film was thinner. Surface temperature of the sphere was measured. Salt additives increased the quenching (vapor film collapse) temperature, because frequency of direct contact between sphere surface and coolant increased. Quenching temperature rises with increased salt concentration. The quenching temperature, however, approaches a constant value when the slat concentration is close to its saturation concentration. The quenching temperature is well correlated with ion molar concentration, which is a number density of ions, regardless of the type of hydrated salts. © 2008 by ASME.

Experiments were conducted in which a molten copper droplet was released into a pool of water. Spontaneous vapor explosion did not occur when the water temperature was 50°C or higher. Spontaneous vapor explosions, however, occurred at a rate of 70% when water temperature was 20°C A series of high-speed video images explores the triggering process of spontaneous vapor explosions: (1) when a molten copper droplet is released into pool of water, a vapor film forms and separates the copper droplet and the surrounding water that is a boiling film, (2) a filament of molten copper grows from the surface and deforms the vapor film, (3) since the filament of molten copper has a small heat capacity and is rapidly quenched, the vapor film condenses along its surface, and the molten copper droplet around the filament directly contacts water, and finally (4) triggering of spontaneous vapor explosions occurs from the filament to the whole molten copper droplet. When filament growth was observed, it triggered a spontaneous vapor explosion in almost all cases. When filament growth was not observed, spontaneous vapor explosions were not observed and vapor films therefore stably formed around the molten copper droplets. The authors thus conclude that filaments from the molten copper trigger spontaneous vapor explosions in highly subcooled water.

The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel instability, core-wide instability and regional instability of ABWR. A real-time simulation was performed by the modal-point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios and resonance frequencies from dominant poles of transfer function on the basis of the AR method using time series measurement data of a core inlet flow of the facility. By utilizing this method, one can estimate stability in any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the BWR-5 core with one third of MOX fuels installed. The decay ratios and resonance frequencies are in good agreement with the analyses results calculated by design analysis code, ODYSY The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and revealed a sufficiently large stability margin even under hypothetical conditions of power level.

Small-scale experiments have been conducted to investigate the triggering mechanism of vapor explosions. In order to attain good repeatability and visibility, a smooth round water droplet was impinged onto a molten alloy surface. This configuration suppresses premixing events prior to triggering. The effect of the water droplet curvature was found to be negligibly small when the droplet diameter is larger than 4.5 mm. Vapor explosion conditions were identical for the molten tin pool depths ranging from 0.5 to 40 mm. The experimental results and the heat conduction analysis suggest that the length scale required for atomizing and fine mixing in the triggering event of the vapor explosion are sufficiently smaller than the molten tin pool depth of 0.5 mm. Six different kinds of materials were used as the pool liquid. The lower limit of the contact temperature in the vapor explosion region closely agrees with the spontaneous nucleation temperature of water. The upper limit of the initial molten alloy temperature decreases when an oxide layer forms on the surface causing an increase of the emissivity of thermal radiation that has a stabilizing effect on the vapor film. When an oxide layer formed on the surface, a water droplet was occasionally entrapped into a molten alloy dome, since the oxide layer prevents the droplet from evaporating coherently. The vapor explosion region obtained for the mirror surface is a conservative estimate, since that for the oxide surface fell into the internal region of mirror surface. Copyright © 2007 by ASME.

To investigate the stability of a boiling water reactor (BWR), the SIRIUS-F facility was designed and built for highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermal hydraulics-neutronics instabilities of the BWR. By using two sets of measured void-fraction distributions in a reactor core section of the SIRIUS-F facility, a real-time void-reactivity feedback simulation was performed on the basis of the modal point kinetics of reactor neutronics and fuel rod thermal conduction. A noise analysis method was performed to calculate decay ratios and resonance frequencies from dominant poles of transfer function based on the AR method using time-series measurement data of a core inlet flow of the facility. Channel and regional stability experiments were conducted for a wide range of operating conditions, including maximum power points along the minimum pump speed line and the natural circulation line of advanced BWR plants. The experimentally obtained decay ratios and resonance frequencies are in good agreement with those calculated by the linear stability analysis code ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics as a function of power and revealed a sufficiently large stability margin even under hypothetical power level conditions.

Japanese PWR utilities desire to employ long-term fuel cycle and high bum-up operations. Corrosion product (crud) deposition on fuel cladding surface would become significant issue in Japanese PWRs, because the large amounts of crud deposition has led to an increase of the dose rate in the primary coolant system and become a root cause of axial offset anomalies (AOA) for other utilities. In order to clarify the contribution factors of the deposition, the effects of steaming rate, pHi-, and boron contents in the test solution on the crud deposition were investigated under sub-cooled boiling and non-irradiated condition. From the test results, it was revealed that the spinel oxide layer was formed on the fuel cladding surface in a simulated Japanese PWR fuel cycle chemistry at 325°C. The oxide layer was easily formed on the heated surface of the fuel cladding, and the oxide thickness was increased with increase of steaming rate and decrease of pHT.

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for heat conduction in a fuel-rod and modal point kinetics of reactor neutronics using measured void fractions in reactor core sections of the thermal-hydraulic loop. In order to estimate decay ratios, an auto-regressive method has been successfully applied for time series data of the core inlet flow rate. Experiments were conducted with the SIRIUS-N facility for the rated operating condition of 3.13GWt natural circulation BWR. Channel and regional stability decay ratios were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to evaluate the stability sensitivity of design parameters such as the power profile, void reactivity coefficients, core inlet sub-cooling, and the fuel rod time constant. © 2005 Taylor & Francis Group, Ltd.

Improving the limit of boiling heat transfer or critical heat flux requires that the cooling liquid can contact the heating surface, or a high wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, contact angle, an indicator of macroscopic wettability, of a water droplet on metal oxide at room temperature was measured by image processing of the images obtained by a CCD video camera. The results showed that the surface wettability on metal oxide pieces of titanium, zircaloy No. 4, SUS-304, and copper was improved significantly by the radiation induced surface activation (RISA) phenomenon. To delineate the effect of RISA on heat transferring phenomena, the Leidenfrost condition and quenching of metal oxides irradiated by y-rays were investigated in this study. In the Leidenfrost experiment, when the temperature of the heating surface reached the wetting limit temperature, water-solid contact vanished because a stable vapor film existed between the droplet and the metal surface; i.e., a Leidenfrost condition obtained. The wetting limit temperature increased with integrated irradiation dose. After irradiation, the wet length and the duration of contact increased, and the contact angle decreased. In the quenching test, high surface wettability, or a highly hydrophilic condition, of a simulated fuel rod made of SUS was achieved, and the quenching velocities were increased up to 20-30% after 300 kGy 60Co γ-ray irradiation. © 2004 Elsevier Inc. All rights reserved.

Experiments were conducted to investigate two-phase flow instabilities due to flashing in a boiling natural circulation loop with a chimney at low pressure. The SIRIUS-N facility was designed to have non-dimensional values nearly equal to those of typical natural circulation boiling water reactor (BWR). The observed instability is suggested to be flashing-induced density wave oscillations, since the oscillation period correlated well with the passing time of single-phase liquid in the chimney section regardless of system pressure, heat flux, and inlet subcooling. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 0.1, 0.2, 0.35, and 0.5 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before heating according to the stability map. © 2005 Elsevier B.V. All rights reserved.

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations. © 2004 Elsevier B.V. All rights reserved.

Experiments were conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney at high pressure. The SIRIUS-N facility was designed to have non-dimensional values which are nearly equal to those of a typical natural circulation BWR. The observed oscillations are found to be density wave oscillations, since the void fractions in the chimney inlet and exit are out of phase. They belong to the Type-I category, since they occur at low flow qualities, according to the Fukuda—Kobori's classification. Moreover, the oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. Two distinct phenomena are found in relation between the oscillation period and liquid passing time in the chimney, indicating that the driving mechanisms of the instabilities are different between low and high pressures. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 1, 2, 4, and 7.2 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarges with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing the control rods. The obtained stability map demonstrates that the nominal operating condition of the ESBWR has a significant stability margin to the unstable region. © 2005 Taylor & Francis Group, Ltd.

Effects of mitigation and suppression of vapor explosion have been investigated by adding a surfactant, polymer, and neutral salt into the water droplet impinging onto a molten alloy pool surface. This configuration was selected to attain good reproducibility and visibility, because premixing events prior to the triggering restrained. In addition, an adequate amount of a surfactant can be adsorbed at the vapor/liquid interface compared to that in the case of conventional configurations such as a molten alloy droplet injecting into a water pool. Dilute anionic and nonionic surfactant aqueoues solutions did not affect the triggering conditions or the pressure pulse generated by vapor explosion, even if the surfactant was added up to a density 25 times higher than the critical micelle concentration. Polymeric additives, which increase the viscosity of a fluid, have little suppression effect on vapor explosion. Spontaneous vapor explosion was, however, suppressed by a 200 wppm polyethylene glycol (molecular weight of 4 x 106) solution, since deposition of the solute due to cloudy-point phenomenon may stabilize the vapor film and prevent the solution from mixing finely. In order to exert this effect, molecular weight should be heavier so that a cloudy-point temperature is below the boiling point at the tested system pressure. Additionally, this threshold concentration became denser as the impingement velocity increased. Thus, a denser concentration and heavier molecular weight should be used to suppress vapor explosion when the vapor film may be subjected to large inertia and/or external force. When a neutral salt was added, the initial molten alloy temperature range where vapor explosion occurred shifted to a higher temperature and became wider. Vapor explosion was observed on the molten zinc surface, which does not explode spontaneously for water. © 2002 Elsevier Science Inc. All rights reserved.

In order to investigate the effect of considering liquid density dependence on local fluid temperature in the thermal-hydraulic stability, a linear stability analysis is performed for a boiling natural circulation loop with an adiabatic riser. Type-I and Type-II instabilities were to investigate according to Fukuda-Kobori's classification. Type-I instability is dominant when the flow quality is low, while Type-II instability is relevant at high flow quality. Type-II instability is well known as the typical density wave oscillation. Neglecting liquid density differences yields estimates of Type-II instability margins that are too small, due to both a change in system-dynamics features and in the operational point. On the other hand, neglecting liquid density differences yields estimates of Type-I stability margins that are too large, especially due to a change in the operational point. Neglecting density differences is thus non-conservative in this case. Therefore, it is highly recommended to include liquid density dependence on the fluid subcooling in the stability analysis if a flow loop with an adiabatic riser is operated under the condition of low flow quality. © 2002 Taylor and Francis Group, LTD.

Experiments have been conducted to investigate an effect of inlet restriction on the thermal-hydraulic stability. A Test facility used in this study was designed and constructed to have non-dimensional values that are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation at the stability boundary was described as a function of heat flux and inlet subcooling independent of inlet restriction. In order to extend experimental database regarding thermal-hydraulic stability to different inlet restriction, numerical analysis was carried out based on the homogeneous flow model. Stability maps in reference to the core inlet subcooling and heat flux were presented for various inlet restrictions using the above-mentioned function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux.

An analytical study is presented on the thermo-hydraulic stability of a boiling natural circulation loop with a chimney at low pressure start-up. The effect of flashing induced by the pressure drop in the channel and the chimney due to gravity head on the instability is considered. A method to analyze linear stability is developed, in which a drift-flux model is used. The analytical result of a stability map agrees very well with the experimental one obtained in a previous report. Instability does not occur when the heater power is too low to generate voids in the chimney and only natural circulation of single phase can be induced. Instability tends to occur when boiling occurs only near the chimney exit due to flashing. This instability phenomenon has some similarities with density wave oscillation, such as the phase difference of temperature between the boiling region and non-boiling region, and the oscillation period which is near to the time required for fluid to pass through the chimney. However, there are also some differences from density wave oscillation, such as the boiling region is very short, and pressure fluctuation can affect void fraction fluctuation.

The compact reversed shear tokamak CREST is a cost competitive reactor concept based on a reversed shear high β plasma and water cooled ferritic steel components. The moderate aspect ratio A = 3.4 and the elongation κ = 2.0 of CREST are very similar to the case of the ITER advanced mode plasma. Presentation of such a concept based on the ITER project should be worth while for formulating a fusion development strategy. The achievement of a competitive cost of electricity (COE) is the first priority for electric power industries. High β and high thermal efficiency are the most effective parameters for achieving a competitive COE. In order to achieve a high efficiency power plant, a superheated steam cycle has been adopted which permits a high thermal efficiency (η = 41%). Current profile control and high speed plasma rotation by neutral beam current drive stabilize the ideal MHD activity up to the Troyon coefficient βN = 5.5. A cost assessment has shown that CREST could generate about 1.16 GW(e) electric power at a competitive cost.

In order to investigate the triggering mechanism of vapor explosion, small-scale experiments have been conducted to observe and examine explosive fragmentation characteristics due to thermal interaction between an impinging droplet and a molten alloy pool. The process that initiates vapor explosion was clearly observed in high-speed videotaping frames: a hemispherical water droplet fragmented from the bottom surface where spontaneous bubble-nucleation took place immediately after making direct contact with a molten alloy pool surface, and small particles of the alloy also fragmented and dispersed due to the high pressure that resulted from vapor explosion. Conditions in which vapor explosion occurred were investigated for various initial temperatures of the droplet and molten alloy peel. It is found from wave height for deformed molten alloy surface that explosion intensity diminishes with decreasing droplet subcooling.

Instability of a boiling natural circulation loop with a chimney at low pressure and low heater power was investigated by linear stability analysis and experiment. A homogeneous and thermodynamic equilibrium model for two-phase flow was used. The effect of flashing induced by pressure drop in the heated channels and the chimmey was considered. The effects of coupling between two boiling channels were investigated. It was found that in-phase-mode instability was apt to occur when channel inlet subcooling was large and boiling began in the chimney. In-phase-mode instability easily occurred when channel length was shortened and chimney length was increased. Out-of-phase-mode instability was apt to occur when chimney length was decreased and boiling began in the channel.

As a candidate for an innovative system generator for fast-breeder reactors, a heat exchanger with direct-contact heat transfer between a melted alloy and water was proposed. In this study, the effect of pressure on the heat-transfer characteristics and the required degree of superheating of the melted alloy above water-saturation temperatures are evaluated during the direct-contact heat-transfer experiment by injecting water into Wood's alloy. As a result of the experiment, the product of the degree of Wood's alloy superheating above the water-saturation temperature and the depth of the feed-water injection point is constant for each pressure. This constant increases as the pressure rises.

Experiments were conducted to investigate two-phase instabilities in a boiling natural circulation loop induced by flashing in a chimney at lower pressure. The type of instability that occurred in the experiments is suggested to be density wave oscillations induced by flashing in the chimney. The differences from other instabilities such as geysering, flow pattern transition instability, and natural circulation oscillations are discussed on the basis of the dynamic characteristics, the oscillation period, and the transient flow resume.